L-MT-09-044, Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (Emcb) Requests for Additional Information (Rais) Dated March 28, 2009

Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (Emcb) Requests for Additional Information (Rais) Dated March 28, 2009

ML092390332
Person / Time
Site:	Monticello
Issue date:	08/21/2009
From:	O'Connor T Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML092390341	List: L-MT-09-044, Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (Emcb) Requests for Additional Information (Rais) Dated March 28, 2009 ML092390333
References
L-MT-09-044, TAC MD9990
Download: ML092390332 (58)

Text

WITHHOLD ENCLOSURE 3 FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 and 9.17 August 21, 2009 L-MT-09-044 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License License No. DPR-22 Monticello Extended Power Uprate: Response to NRC Mechanical and Civil Engineering Review Branch (EMCB) Requests for Additional Information (RAIs) dated March 28, 2009 (TAC MD9990)

References:

1. NSPM letter to NRC, License Amendment Request: Extended Power Uprate (L-MT-08-052) dated November 5, 2008, (TAC MD9990)

Accession No. ML083230111

2. Email P. Tam (NRC) to G. Salamon, K. Pointer (NSPM) dated March 28, 2009, Monticello - Draft RAIs from Mechanical & Civil engineering Branch re: proposed EPU amendment (TAC MD9990)

Accession No. ML090880002 Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota corporation (NSPM), requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications to increase the maximum authorized power level from 1775 megawatts thermal (MWt) to 2004 MWt.

On March 28, 2009, the U.S. Nuclear Regulatory Commission (NRC) Mechanical and Civil Engineering Review Branch (EMCB) provided the requests for additional information (RAIs) contained in Reference 2. Enclosure 1 provides the proprietary response to EMCB RAIs in References 2. A non proprietary version of Enclosure 1 is contained in Enclosure 3. GEH requests this proprietary information to be withheld from public disclosure in accordance with 10 CFR 2.390(a)4 and 9.17(A)4. An affidavit supporting this request is provided in . Enclosure 4 is provided for information.

In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Minnesota Official without the proprietary version.

Monticello Nuclear Generating Plant 2807 West County Road 75

• Monticello MN 55362

Document Control Desk L-MT-09-044 Page 2 of 2 Summary of Commitments

1. Confirmation that Feedwater and Condensate pump and heater replacement modifications are complete and meet the code allowables will be provided to the NRC prior to implementation of the EPU license amendment request.

2. Confirmation that modification of support W H - 1 4 3 is complete will be provided to the NRC prior to implementation of the EPU license amendment request.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on August g, 2009.

Timothy J. OIConnor Site Vice President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

ENCLOSURE 3 NSPM RESPONSE TO EMCB RAIs DATED MARCH 28, 2009 Non Proprietary

L-MT-09-044 Non Proprietary Page 1 of 46 EMCB RAI No. 1 Provide a table which contains information on plant operating parameters similar to Table 1-2 and include a column for OLTP. Include design and maximum temperatures for reactor recirculation system (RRS) vessel outlet and inlet nozzles and feedwater (FW) nozzles.

NSPM RESPONSE Plant Operating Conditions OLTP CLTP1 EPU Thermal Power (MWt) 1670 1775 2004 Vessel Steam Flow (Mlb/hr) 6.78 7.26 8.34 Full Power Core Flow Range Mlb/hr 43.2 to 60.5 47.5 to 60.5 57.0 to 60.5

% Rated 75 to 105 82.4 to 105 99.0 to 105 Maximum Normal Dome Pressure 1025 No Change No Change (psia)

Maximum Normal Dome 548 No Change No Change Temperature (ºF)

Pressure Upstream of TSV (psia) 965 970 952 Full Power Feedwater Flow (Mlb/hr) 6.75 7.24 8.31 Temperature (ºF) 377 383.0 395.8 Core Inlet Enthalpy (Btu/lb)2 524.6 523.7 523.0

1. Based on current reactor heat balance; 2. At 100% core flow condition Reactor Nozzle OLTP CLTP EPU Value RRS Outlet Design Temperature 575°F No Change No Change 1

RRS Outlet Maximum Temperature 546°F 549°F 548°F RRS Inlet Design Temperature 575°F No Change No Change RRS Inlet Maximum Temperature1 546°F 549°F 548°F FW Nozzle Design Temperature 575°F No Change No Change FW Nozzle Maximum Temperature 376°F 385°F 398°F

1. Maximum temperature is saturation temperature for reactor with no feedwater flow assumed. OLTP value is based on normal reactor pressure of 1000 psig, CLTP value is based on normal reactor pressure of 1025 psig and EPU value is based on normal reactor pressure of 1010 psig.

L-MT-09-044 Non Proprietary Page 2 of 46 EMCB RAI No. 2 Confirm whether the current licensing basis criteria for high energy line break (HELB) are the criteria contained in the Giambusso/Schwencer letters (1972-73).

NSPM RESPONSE These criteria were not changed for EPU. USAR Appendix I,Section I.1, defines the evaluation criteria for HELBs. The USAR states:

The criteria used for the determination of the high energy lines and the effects of the postulated breaks on these lines on safe shutdown equipment are the December 18, 1972 Giambusso letter (Reference 2) as clarified by Standard Review Plan (SRP) 3.6.1 (Reference 3), SRP 3.6.2 (Reference 4), and Generic Letter 87-11 (Reference 21).

These criteria are utilized as the basis for the determination of the high energy lines, break locations, and the evaluation of effects on Safe Shutdown (SSD) equipment.

The associated USAR references are:

2. Letter from A. Giambusso, Deputy Director for Reactor Projects, to Northern States Power Company,

Subject:

High Energy Breaks Outside of the Containment, December 18, 1972.

3. Standard Review Plan 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, Rev. 1, July 1981.

4. Standard Review Plan 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Rev. 1, July 1981.

21. NRC (F J Miraglia) Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, June 19, 1987.

Staff review and acceptance of the analyses performed and the measures taken in response to the December, 1972 letter from A Giambusso is documented in the July 29, 1974 letter from Karl R Goller to Northern States Power Co. Letter, AEC to NSP, United States Atomic Energy Commission - Safety Evaluation by the Directorate of Licensing, Docket No. 50-263, Monticello Nuclear Generating Plant - Analysis of the Consequences of High Energy Piping Failures Outside Containment Documentation of further Staff review is provided in letter, NRC to NSP, Monticello -

High Energy Line Break Analysis (TAC No. 61788), June 13, 1990.

L-MT-09-044 Non Proprietary Page 3 of 46 EMCB RAI No. 3(a)

PUSAR Section 2.2.1 states that corrective actions are underway to perform HELB analysis upgrades at Monticello due to changes in pipe break methodology.

Explain why corrective actions are in place to upgrade the Monticello pipe break methodology.

NSPM RESPONSE PUSAR Section 2.2.1 states:

Technical Evaluation No changes to the implementation of the existing criteria for defining pipe break and crack locations and configurations are being made for EPU . . .

Changes in Methods of Analysis The results provided for HELB events affected by EPU, specifically, the liquid line breaks in the Feedwater, Condensate, and RWCU systems show much larger changes than would be expected due to the small changes in pump discharge pressures and small enthalpy changes as a result of EPU. The results are driven by conservative changes in analysis methods resulting from corrective actions underway to perform HELB analysis upgrades at Monticello.

The criteria used to determine high energy lines has not changed with EPU, see RAI 2 above. The changes NSPM referred to are covered in corrective action program action request AR01131913, HELB Program Documentation Deficiencies, which documents a summary of issues being addressed. The most significant changes are related to the assumptions used in determining mass and energy releases from postulated breaks and upgrade of the computer code from GOTHIC version 4.0. The EPU liquid break calculation inputs have been upgraded to consider:

1. Double-ended break flow to include flow from both ends of postulated breaks

2. System depletion to include mass and energy that exists in system piping and pressure vessels

3. A conservative change in assumption for isolation valve stroke time from ASME Section XI Limiting Stroke time to the value listed as the maximum valve operating time in the USAR. If break detection logic exists, valve stroke is initiated when the logic detects the break.

4. A conservative change for flow reduction assumptions with valve closure. CLTP analysis assumed flow was reduced proportional to isolation valve percent closed position. The EPU analysis assumed 100% break flow until isolation valve was fully closed.

L-MT-09-044 Non Proprietary Page 4 of 46

5. The liquid mass from fire protection sprinkler systems postulated to actuate from HELB events was included

6. Upgrade computer code from GOTHIC version 4.0 to GOTHIC version 7.1 or later versions The assumption changes noted above are based on recommendations from site self assessments. These changes will bring the HELB program into closer alignment with industry standards and correct identified deficiencies. The failure to consider fire protection sprinkler system actuation for appropriate HELBs resulted in the issuance of LER 2008-001, Non-Conservative HELB Analysis discovered during EPU, and is documented under AR1125675.

Re-analysis of all HELB breaks and an evaluation of affected EQ components have been completed; formal updating of EQ program documents are the only actions remaining. These actions are being performed coincident with EQ program updates required by EPU.

EMCB RAI No. 3(b)

Verify whether the Monticello pipe break methodology upgrade is based on SRP Section 3.6.2, MEB 3-1 criteria. If not, provide supporting justification.

NSPM RESPONSE:

As noted above in the response to Part a) of this question, there is no change to the pipe break methodology at Monticello. The changes involve a re-analysis of breaks using more conservative assumptions of mass and energy release.

L-MT-09-044 Non Proprietary Page 5 of 46 EMCB RAI No. 4 ELTR 1 and ELTR 2 both recommend that HELB evaluations for High Pressure Core Spray (HPCS) and Building Heating Line be performed on plant-specific power uprate submittals. Please indicate where in the proposed LAR submittal these evaluations have been performed or provide the HELB plant-specific evaluations for these systems at EPU conditions.

NSPM RESPONSE Monticello does not have a High Pressure Core Spray system, see USAR Section 6.2.

The criteria for HELB consideration at Monticello are for piping systems that are >275 psig and >200°F, see USAR Appendix I.2. This is based on United States Atomic Energy Commission - Safety Evaluation by the Directorate of Licensing, Docket No. 50-263, Monticello Nuclear Generating Plant - Analysis of the Consequences of High Energy Piping Failures Outside Containment, July 29, 1974 (Enclosure 4). Building heating lines at Monticello do not meet criteria for consideration under the HELB program and therefore were not evaluated.

L-MT-09-044 Non Proprietary Page 6 of 46 EMCB RAI No. 5 Page 2-23 states that:

During the 6.3 percent rerate in 1996, only one new case was reanalyzed at CLTP for the RWCU system - a break in the system suction piping at the outboard isolation valve. For this reason a detailed comparison of CLTP and EPU results for HELBs in the RWCU system is not possible.

The statement that, For this reason a detailed comparison of CLTP and EPU results for HELBs in the RWCU system is not possible is not clear. Please provide clarification.

NSPM RESPONSE The CLTP analysis of RWCU HELBs evaluated the terminal end break and crack case at the inlet to the RWCU heat exchanger. The evaluation used the mass and energy release rates for a break just outboard of the outboard isolation valve. These were considered the bounding cases and other cases were not run. For EPU, eight HELB locations, covering all possible breaks and cracks, were evaluated.

The response to RAI 3.a above explains changes in assumptions used in evaluation of the EPU HELB cases. As noted on PUSAR page 2-21:

Because of these changes in methodology, a comparison of the results between EPU and CLTP conditions shows a significantly larger change than would normally be expected based on the small changes in process fluid temperatures and enthalpy resulting from EPU based on previous industry experience.

Monticello has chosen not to perform a full re-analysis of these specific liquid line HELBs at CLTP conditions because it was determined that our effort should be focused on completing the corrective actions using bounding conditions. Thus, a detailed breakdown of the magnitude of the change is caused by EPU versus the change resulting from the changes in methods and correction of errors is not provided.

A comparison of the results between EPU and CLTP conditions was not done since it would have required the creation of an additional 12 calculations to define CLTP conditions with the new assumptions included. This significant effort was not warranted as the bounding analysis completed for EPU have addressed the desired CLTP analysis improvements. Results of a comparison between the single CLTP RWCU HELB case and the similar EPU HELB case is discussed in RAI 6 below. Re-analysis of all HELB breaks and an evaluation of affected EQ components have been completed; formal updating of EQ program documents are the only actions remaining. These actions are being performed coincident with EQ program updates required by EPU.

L-MT-09-044 Non Proprietary Page 7 of 46 EMCB RAI No. 6(a)

The same paragraph on page 3-23, as above, in reference to the reactor water cleanup (RWCU), continues as follows:

For the break location that was analyzed during Rerate, new mass and energy release calculations considered additional blowdown sources that had not been considered in the previous 1996 analysis. This resulted in an increase in integrated mass release of about 90% and an increase in integrated energy release of 63 percent.

Confirm that the 90% and 63% increases are referring to the proposed EPU.

NSPM RESPONSE:

The 90% and 63% increases are not referring to the proposed EPU. It is referring to the change in assumptions as noted in response to RAI 3 above rather than system operating condition changes resulting from EPU.

If the CLTP HELB cases were run using similar assumptions, the changes in mass and energy releases would be minor as a result of EPU.

As noted on PUSAR page 2-21:

A review of the results from several recent EPU submittals concluded that, in most cases, environmental conditions are bounded by previous analyses, confirming that EPU produces relatively minor effects.

EMCB RAI No. 6(b)

Please explain how the effects of the increased mass and energy release have been evaluated, include evaluations of pipe whip restraints and jet targets.

NSPM RESPONSE Changes in mass and energy were evaluated for impacts on HELBs using the GOTHIC code. This allowed a determination of time histories for all plant areas to evaluate effects on temperature, pressure and flooding. Differential pressures between plant areas verified acceptable margins for structures such as block walls. The effects of changes to temperature, pressure and flooding have been evaluated for impact on the

L-MT-09-044 Enclosure 3 Non Proprietary Page 8 of 46 environmental qualification (EQ) of equipment. Upgrades to EQ files to document this evaluation are in progress.

RWCU pipe whip, jet impingement and safe shut down analyses following postulated pipe breaks or cracks are provided in USAR Appendix I. The RWCU high energy lines are located in the RWCU compartment, steam chase; MG set room, and the North West side of elevations 962 and 935 of the reactor building. There are no postulated breaks in the MG set room and the reactor building elevations 962 and 935 based on seismic analysis. There are no pipe whip targets for the RWCU piping in the steam chase.

The safe shutdown evaluation for the RWCU compartment in Appendix I does not rely on pipe whip restraints or jet impingement shields to protect any equipment or structures.

The effects of pipe whip and jet impingement in this area do not result in the loss of components required to mitigate the break and shut down the reactor. Therefore there is no impact on RWCU pipe whip and jet impingement due to EPU.

EMCB RAI No. 7 Page 2-37 states that: The combination of stresses was evaluated to meet the requirements of the pipe break criteria. Based on these criteria, no new postulated pipe break locations were identified. For systems affected by the EPU, specifically steam (all EPU affected steam lines) and FW lines (including condensate), provide a pipe break analysis summary table (that includes the main steam increased turbine stop valve (TSV) closure transient loads in the analysis) which compares values at EPU and CLTP conditions and shows code equation stresses and CUFs compared to break limit for stresses and CUFs. Include pipe break locations and types selected for CLTP and EPU. Include lines inside and outside containment.

NSPM RESPONSE Systems that have piping meeting the MNGP design basis criteria for classification as High Energy include Main Steam, Condensate, Feedwater, Residual Heat Removal (RHR), Core Spray (CS), High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), Reactor Water Cleanup (RWCU), Off Gas, Control Rod Drive (CRD), Zinc Oxide Injection (GEZIP), and Standby Liquid Control (SLC). The parameters used for stress analysis in the high energy portions of these systems are unchanged due to EPU except in the Main Steam, Condensate, Feedwater, and GEZIP systems.

The Main Steam system analysis results including TSV closure loads are provided in the table below. The stress result for the Main Steam location with the maximum HELB break postulation equation result is also included in the table. The stress at that

L-MT-09-044 Non Proprietary Page 9 of 46 location does not meet (is less than) the current design basis criteria to require a postulated break. Hence, there is no Main Steam break outside containment postulated based on stress criteria. Other postulated break locations are based on configuration (e.g., terminal ends) which is not changed by EPU. Note that in the current design basis, specific HELB locations are not postulated inside containment. The current design basis does not include fatigue analysis of the Main Steam piping. Due to the revised analysis of the turbine stop valve closure loads, comparison to pre-EPU values is not meaningful.

The Main Steam evaluation results shown below are performed for the EPU pressure, temperature and flow parameters, including the TSV closure loads.

Main Steam Outside Containment - Maximum EPU Results (Highest Interaction Ratio):

Load Service Node Stress Allowable Ratio Combination Level psi psi S/Allow P+DW B X7A 6877 15000 0.46 TH Range B TURB 19441 22500 0.86 P+DW+TSV B 268 12236 18000 0.68 DW+TSV+SRV+SSE D 268 13795 26325 0.52 HELB DW+TH+OBE B TURB 27559 30000 0.92 The maximum Feedwater system operating temperature is 397.7oF at EPU conditions for the Feedwater piping from the outboard containment isolation valve to the containment and inside containment. This value is bounded by the original analysis temperature of 400oF. The design pressure for this portion of the Feedwater system is unchanged by EPU. Therefore this piping is unaffected by EPU relative to HELB postulation.

The feedwater piping and condensate piping from the condensate pump suction to the containment isolation valves will be re-analyzed during the Feedwater and Condensate pump and heater replacement modification process. High Energy Line Breaks and pipe whip restraints in the high energy portion of this piping will be evaluated at that time.

GEZIP connections to the portion of the Feedwater system will be analyzed as part of the modification process. Details of the modifications to this piping are not yet finalized.

The design will maintain stresses in the condensate and FW piping within code allowable limits of ANSI-B31.1-1977, including Winter 1978 Addenda and the requirements of USAR Chapter 12 including USAR Appendix I. Confirmation that the modifications are complete and meet the code allowables will be provided to the NRC in a separate letter. The FW and condensate system modifications are scheduled for completion during RFO25 in 2011.

L-MT-09-044 Non Proprietary Page 10 of 46 EMCB RAI No. 8 , PUSAR Section 2.2.1.2, Liquid Line Breaks, on page 2-23 states that:

The mass and energy releases for HELBs in the RWCU, FW, Condensate, CRD, Standby Liquid Control, and Zinc Injection (GEZIP) systems and instrument and sample

{3}

lines may be affected by EPU and were re-evaluated at EPU conditions. (( ))

evaluations of liquid line breaks have been performed at EPU conditions.

Provide similar summaries as in RAI 7 for the RWCU line breaks at EPU conditions.

NSPM RESPONSE From a HELB postulation viewpoint, there is no change in RWCU piping analysis temperature or design pressure due to EPU. Consequently, the pipe break postulation stress evaluations for RWCU are not changed at EPU conditions. Changes in mass and energy release are primarily due to the change in assumptions identified in response to RAI 3a above.

EMCB RAI No. 9 Indicate whether the FW lines have been structurally analyzed for any flow instabilities and loads due to water hammer or other flow transients and whether reanalysis has considered the EPU higher flows for these transients in evaluating pipe stresses, pipe breaks and pipe supports.

NSPM RESPONSE The current analysis of the FW lines contains no structural analysis for any flow instabilities or loads due to water hammer or other flow transients. Such analysis was not performed at EPU conditions.

L-MT-09-044 Enclosure 3 Non Proprietary Page 11 of 46 EMCB RAI No. 10 Are there any new liquid or steam line pipe break locations that need to be postulated due to EPU conditions?

NSPM Response There are no new liquid or steam line pipe break locations that need to be postulated due to the change in process conditions at EPU.

Systems that are reconfigured by plant modifications (e.g., condensate and feedwater piping as identified in response to RAI 7) are evaluated during the modification process for HELB break postulation.

EMCB RAI No. 11(a)

For main steam (MS) and FW piping, state the design basis (DB) code for Class I and Class II piping and pipe supports.

EMCB RAI No. 11(b)

Verify that all structural evaluations of SSCs, required for EPU, were performed in accordance with the DB codes of record for piping and pipe supports. If a different code than the DB code of record was used, provide a justification.

NSPM Response

a. The MS piping system and associated branch piping (inside containment) were evaluated for compliance with the ASME Section III, Division I, 1977 Edition with Addenda up to and including Winter 1978 Piping Code stress criteria, including the effects of EPU on piping stress, piping supports including the associated building structure, piping interfaces with the RPV nozzles, containment penetrations, flanges, and valves. The requirements of ANSI B31.1-1977 through the W1978 addenda are used for FW piping and supports.

b. All structural evaluations of SSCs, required for EPU, were performed in accordance with the DB codes of record for piping and pipe supports as indicated on page 2-36 of the PUSAR.

L-MT-09-044 Non Proprietary Page 12 of 46 EMCB RAI No. 12 Page 2-36 of the PUSAR states that, The effects of the EPU conditions have been evaluated for the following piping [BOP] systems: A list of piping systems follows this statement. On page 2-37 of the PUSAR, it is stated that, These piping systems have been evaluated using the process defined in Appendix K of ELTR1 and found to meet the appropriate code criteria for the EPU conditions, when in fact evaluations of many of these systems, including RHR and MS, has not been completed, as shown by the submitted EPU LAR, see PUSAR Table 2.2-2d. In addition, Enclosure 8, Table 8-2 states that EPU planned modifications include, Revise documentation to incorporate revised pressure and temperature ratings for specific piping systems affected by EPU. Modify supports as required by the analyses.

EMCB RAI No. 12(a)

The above PUSAR statements are not consistent. Please clarify the apparent inconsistency.

EMCB RAI No. 12(b)

The proposed EPU LAR indicates that some EPU evaluations have not been completed for the staff to review. The acceptability of the proposed EPU LAR will be determined based upon the results of the LAR evaluation reviews that are performed by the staff in accordance with the policies and procedures set forth in LIC-101, License Amendment Review Procedures. Please provide a schedule of completion of these analyses and submittal of your evaluation results which shows that piping and pipe supports meet code allowable. Also, submit a schedule of completion for EPU required piping and pipe support modifications.

NSPM Response Response to Part a The referenced statement on page 2-37 indicating that pipe systems meet code requirements is intended to apply to piping stresses. Later on the same page, under the heading of Pipe Supports, the structures listed on Table 2.2-2d are discussed.

Based on the ELTR1 Appendix K methodology, the components listed were found to exceed code limits. Further, more detailed analysis may resolve some of these issues; others may require modification. This is consistent with the referenced statement from EPU LAR Enclosure 8 which indicates supports being modified as required by analysis.

L-MT-09-044 Non Proprietary Page 13 of 46 Response to Part b All piping and support evaluations required in ELTR1 have been completed using the methodology of Appendix K or by a more detailed analytical method.

Completion of piping support detailed analysis and/or modifications for items listed in Table 2.2-2d was scheduled for the 2009 outage RF024. The current status of work shown on PUSAR Table 2.2-2d is provided below:

Table 2.2-2d Piping Components Requiring Further Reconciliation Item System Current Status 1 Main Steam (Outside Containment) Refined analysis is complete, all piping components and supports meet code allowables.

2 Feedwater and Condensate (from Replacement of feedwater heaters, condensate pump to the feedwater MO condensate and feedwater pumps will valves downstream of the HP Heaters), result in nozzle changes that will due to pending pump changes impact piping layout and analysis.

Final design of these components is still in progress and is scheduled for completion in the 2011 refueling outage. NSPM will complete piping analysis and modifications as noted in response to RAI 7.

3 Torus Attached Piping Refined analysis is complete, all modifications are complete with exception of one support, TWH-143, which will be completed on-line prior to implementation.

Confirmation that modification of support TWH-143 is complete will be provided to the NRC prior to implementation of the EPU license amendment request.

4 RHR (BOP Condensate Service Water Refined analysis is complete, all piping Lines) components and supports meet code allowables.

5 Cross Around Piping Replacement of CARVs and CARV discharge piping during the RFO impacted this analysis. Prompt evaluations of field changes from this work are complete and all piping and supports meet code allowables.

L-MT-09-044 Non Proprietary Page 14 of 46 Item System Current Status 6 CARV Discharge Piping Replacement of CARVs and CARV discharge piping during the RFO impacted this analysis. Prompt evaluations of field changes from this work are complete, all piping and supports meet code allowables.

EMCB RAI No. 13 a) Provide a list of systems (inside and outside containment) for which temperature, pressure, flow and mechanical loads have been increased due to EPU. Please include OLTP and EPU values.

b) Provide a brief summary that shows the EPU maximum code equation stresses compared to CLTP for the affected systems. For MS, FW and condensate see RAI 17, below. Include fatigue evaluation CUFs, where applicable. It is noted, that although the tables in Section 2.2 of the PUSAR include, for some BOP systems, the percentage increases for pipe stresses and pipe support loads, varying from 9 to 72 percent increases, due to temperature or pressure increases, these percentages are not indications that piping and pipe supports meet code equation allowable values, without providing maximum resulted values compared to code allowable.

NSPM Response The system temperature, pressure, and flow changes due to EPU that are not bounded by the parameters used in the existing stress analyses are shown in Table 1, below.

The maximum code equation stresses for Main Steam at EPU conditions are summarized in Table 2, below. The maximum code equation stresses for BOP systems are summarized in Table 3, below. The current design basis does not include fatigue analysis of the Main Steam piping.

L-MT-09-044 Non Proprietary Page 15 of 46 Table 1 MNGP EPU Piping Analysis Input Parameter Changes CLTP EPU Value Item Parameter OLPT Value Inside Containment 1 Main Steam Flow (Lbm/hr) 6.78E+6 7.262E+6 8.524E+06 2 Feedwater, from outboard containment isolation valves (FW-91-1 and FW 2) to RPV Flow (Lbm/hr) 6.83E+6 7.313E+06 8.575E+06 3 Core Spray (CS)

Temperature (F) 180 196.7 212 Outside Containment 1 Main Steam, upstream of TSV Flow (Lbm/hr) 6.78E+6 7.262E+06 8.524E+06 2 Feedwater, From MO-1614/1615 to FW-91-1/FW-91-2 Flow (Lbm/hr) 6.75E+6 7.235E+6 8.497E+06 From pumps to MO-1614 and MO-1615 Temperature (F) 400 400 Note 1 Pressure (psig) 1550 1550 Note 1 Flow (Lbm/hr) 6.75E+6 7.235E+6 8.497E+06 3 Condensate, from Condensate pump suction to Feedwater pump Temperature (F) 302 310 Note 1 Pressure (psig) 434 434 Note 1 Flow (Lbm/hr) 6.75E+6 7.235E+6 8.497E+06 4 Torus Attached Piping (CS, HPCI, RCIC, Note 2)

Temperature (F) 180 196.7 212

L-MT-09-044 Non Proprietary Page 16 of 46 Table 1 MNGP EPU Piping Analysis Input Parameter Changes CLTP EPU Value Item Parameter OLPT Value 5 Emergency Service Water, ECCS Pump Room Ventilation Units (V-AC-4/5) Outlet Lines Temperature (F) 120 120 122 6 Extraction Steam Operating Temperature (F)

To Heater E-11 177 183 186 To Heater E-12 236 246 253 To Heater E-13 313 315 323 To Heater E-14 344 348 358 To Heater E-15 386 396 407 Design pressure (psig)

To Heater E-11 -8 -7 -6 To Heater E-12 8 13 16.8 To Heater E-13 66 68 79 To Heater E-14 111 117 136.4 To Heater E-15 197 220 254 Flow (Mlbm/hr)

To Heater E-11 0.404 0.592 0.700 To Heater E-12 0.371 0.423 0.490 To Heater E-13 0.443 0.444 0.525 To Heater E-14 0.806 0.893 1.164 To Heater E-15 0.388 0.443 0.548 7 Heater Drains Operating Temperature (F)

From Heater E-11 173 180 183 From Heater E-12 236 243 250 From Heater E-13 241 248 254 From Heater E-14 315 318 327 From Heater E-15 343 349 359

L-MT-09-044 Non Proprietary Page 17 of 46 Table 1 MNGP EPU Piping Analysis Input Parameter Changes CLTP EPU Value Item Parameter OLPT Value Design pressure (psig)

From Heater E-11 -8 -7 -6.6 From Heater E-12 7 12 15 From Heater E-13 54 64 74 From Heater E-14 96 110 128 From Heater E-15 184 215 238 Flow (Mlbm/hr)

From Heater E-11 2.52 2.80 3.43 From Heater E-12 2.04 2.20 2.73 From Heater E-13 1.67 1.78 2.24 From Heater E-14 1.22 1.34 1.71 From Heater E-15 0.39 0.44 0.55 8 Service Water Inlet Temperature (F) 85 90 90 9 Cross Around Temperature (F) 387 393 407 Pressure (psig) 197 214 254 Flow (Mlbm/hr) 6.33 6.75 7.91 10 Cross Around Relief Valve Inlet Temperature (F) 381 389 403 Pressure (psig) 182 204 242 Flow 5.66 6.05 7.03 11 Moisture Separator Drain Temperature (F) 383 392 403 Pressure (psig) 202 204 242 Flow (Mlbm/hr) 0.6728 0.7011 0.877 Note: 1. Due to the planned extensive piping modification to the Condensate and Feedwater systems, this piping is analyzed for EPU condition changes as part of the modification process (Reference response to RAI 7).

2 Torus attached RHR piping is currently analyzed at a temperature higher than the peak torus temperature and is therefore not changed by EPU.

L-MT-09-044 Non Proprietary Page 18 of 46 Table 2 MNGP EPU Main Steam Piping and Support Results Summary The Main Steam evaluation results shown below are performed for the EPU pressure, temperature and flow parameters, including the TSV closure loads.

Main Steam Inside Containment - Maximum EPU Results (Highest Interaction Ratio)

Maximum Pipe Stresses Load Service Combination Level Node Stress Allowable Ratio psi psi S/Allow P+DW B 161 7709 15000 0.51 TH Range B 203 22940 22998 1.00 P+DW+OBE B U08 17823 18000 0.99 DW+TSV+SRV+SSE D U08 31261 36000 0.87 Note: 1. High Energy Line Breaks locations are not postulated inside containment.

2. Due to the revised analysis of the turbine stop valve closure loads, comparison to pre-EPU values is not meaningful.

Maximum SRV Flange Loads Inlet Flange Service Node Moment Allowable Ratio Load Condition Level ft-lb ft-lb M/Allow DW + TH B U07 14558 34083 0.427 DW + TH + Level B Dynamic B U07 39362 68167 0.577 DW + TH + Level D Dynamic D U07 65909 99750 0.661 Outlet Flange Service Node Moment Allowable Ratio Load Condition Level ft-lb ft-lb M/Allow DW + TH B U08 13663 31000 0.441 DW + TH + Level B Dynamic B U08 34907 62083 0.562 DW + TH + Level D Dynamic D U08 57547 91250 0.631

L-MT-09-044 Non Proprietary Page 19 of 46 Table 2 MNGP EPU Main Steam Piping and Support Results Summary Main Steam Inside Containment - Maximum EPU Results (Highest Interaction Ratio)

Maximum RPV Nozzle Loads RPV Nozzle N-3D Service Node Fx Fy Fz Mx My Mz Loads Level lb lb lb ft-lb ft-lb ft-lb Maximum Loads B 101 6667 18555 4979 67422 18193 98764 Allowables B 101 19392 51712 19392 258562 32320 258562 Maximum/Allowable B 101 0.344 0.359 0.257 0.261 0.563 0.382 Maximum Flue Head Anchor Loads Penetrations X7A, X7B, X7C, X7D - Side Bolt Evaluation Service Node Tension Shear T allow S allow IR Load Condition Level lb lb lb lb T/Ta+S/Sa DW+TH+SSE+BREAK (X7D) D 22 106702 17509 157500 96250 0.859 DW+TH+SSE+BREAK (X7A) D 30 107227 16683 157500 96250 0.854 Maximum Support Loads MS Relief Valve Discharge Line Support RV25A-H1 (spring hanger)

Max Min Service Node Load Allowable IR Load Allowable IR Load Condition Level lb lb Max/Allow lb lb Allow/Min DW+TH+SRSS(TSV,SRV,OBE) B 285 1341 1344 0.998 1162 780 0.671

L-MT-09-044 Non Proprietary Page 20 of 46 Table 2 MNGP EPU Main Steam Piping and Support Results Summary Main Steam Outside Containment - Maximum EPU Results (Highest Interaction Ratio)

Maximum Pipe Stresses Load Service Node Stress Allowable Ratio Combination Level psi psi S/Allow P+DW B X7A 6877 15000 0.46 TH Range B TURB 19441 22500 0.86 P+DW+TSV B 268 12236 18000 0.68 DW+TSV+SRV+SSE D 268 13795 26325 0.52 HELB DW+TH+OBE B TURB 27559 30000 0.92 Maximum Turbine Loads Load Service Node Mx Allowable Ratio Mz Allowable Ratio Combination Level ft-lb ft-lb Mx/Allow ft-lb ft-lb Mz/Allow DW B

• 32244 413000 0.078 171446 722000 0.237 DW + TH B

• 271321 413000 0.657 302310 722000 0.419

• Note: Loads from all turbine nodes were combined Maximum Support Loads Main Steam Line Support PS-16, Node 283 Max Service Load Allowable IR Load Condition Level Component lb lb Max/Allow DW+TH+SRSS(TSV,SRV,OBE) B Anchor bolt 20026 20731 0.966

L-MT-09-044 Non Proprietary Page 21 of 46 Table 3 MNGP EPU BOP Piping and Support Results Summary Maximum EPU Results (Highest Interaction Ratio)

Extraction Steam Maximum Pipe Stresses ANSI Load B31.1 Heater Stress Allowable Ratio Combination EQ. psi psi S/Allow P+DW 11 15A 7944 15000 0.53 P+DW+OCC 12 15A 7967 18000 0.44 P+DW+TH 14 15A 25599 37500 0.68 Note: OCC represents stresses/loadings from the occasional loadings from simplified seismic analysis using Uniform building code (UBC) methodology.

Maximum Support Loads, Support for Heater 14APS-16, Node 283 Pre-EPU EPU % EPU Load Condition IR Increase IR DW+TH+OCC 0.79 4.64 0.827 Heater Drains & Vents Maximum Pipe Stresses ANSI Load B31.1 Heater Stress Allowable Ratio Combination EQ. psi psi S/Allow P+DW 11 14A-13A 3772 15000 0.25 TH 13 14A-13A 19564 22500 0.87 Maximum Support Loads, Support HDH-73, Feedwater Heater E-13B Dump Line Pre-EPU Increase EPU Support EPU Load, Load Condition Load, lb  % lb Capacity IR DW+TH 1452 51.40% 2198 2200 0.999

L-MT-09-044 Non Proprietary Page 22 of 46 Table 3 MNGP EPU BOP Piping and Support Results Summary Moisture Separator Drain Lines Maximum Pipe Stresses ANSI Load B31.1 Tanks Stress Allowable Ratio Combination EQ. psi psi S/Allow P+DW 11 T6A-T6D 4452 15000 0.40*

TH 13 T6A-T6D 18050 22500 0.80

• Reflect results of prompt evaluation of as-built conditions from CARV modifications completed during the 2009 refueling outage.

Maximum Support Loads, Support CDH-64, Moisture Separator 11 Drain Line Support Pre-EPU EPU % EPU Load Condition IR Increase IR DW+TH 0.64 30 0.83 Core Spray Maximum Pipe Stresses Pre-EPU EPU % EPU Load Condition IR Increase IR TH 0.91 8.9 0.99 Maximum Support Loads, Support TWH-86, Core Spray Pump Discharge Line Support Pre-EPU EPU Pre-EPU EPU Load, Load Condition Load, lb lb IR IR DW+TH+OBE 1471 1480 0.99 0.996

L-MT-09-044 Non Proprietary Page 23 of 46 Table 3 MNGP EPU BOP Piping and Support Results Summary RCIC Injection Maximum Pipe Stresses Pre-EPU EPU % EPU Load Condition IR Increase IR TH 0.83 8.9 0.904 Maximum Support Loads, Support H-1 Pre-EPU EPU % EPU Load Condition IR Increase IR DW+TH+OBE 0.997 0 0.997 NOTE: H-1 not affected by EPU increases. Support is remote from temperature increase HPCI Injection Maximum Pipe Stresses Pre-EPU EPU % EPU Load Condition IR Increase IR DW+TH+OBE 0.78 8.9 0.85 Maximum Support Loads, Support SR-393, Suction Supply Line Support Pre-EPU EPU % EPU Load Condition IR Increase IR DW+TH 0.966 0 0.966 NOTE: SR-393 not affected by EPU increases. Support is remote from temperature increase

L-MT-09-044 Non Proprietary Page 24 of 46 EMCB RAI No. 14 Verify whether the increased flow rate due to EPU affects the structural analysis (pipe stress and support loads) of only the MS and FW piping.

NSPM RESPONSE The current design basis includes fluid transient loads only in the Main Steam system.

The increased Main Steam flow rate due to EPU is included in the structural analysis (pipe stress and support loads) of the Main Steam piping and attached branch lines.

The current licensing basis (refer to USAR Section 12.2.1) does not include flow induced load analyses for the Feedwater piping and none was added for EPU.

EMCB RAI No. 15 The reactor coolant pressure boundary (RCPB) piping systems structural evaluation is contained in Section 2.2.2 of the PUSAR. Please provide structural evaluations for the residual heat removal (RHR) low pressure coolant injection (LPCI) and core spray systems and whether their piping and supports are structurally adequate for the EPU conditions.

NSPM RESPONSE The only system condition change in either RHR (LPCI) and CS is operation with suppression pool water increased from a peak temperature of 196.7oF (current) to 212oF (EPU). The injection portions of these systems near the reactor were originally analyzed at 570oF and are unaffected by this change. The remainder of the RHR (LPCI) system was originally analyzed at a temperature of 330 ºF representing the shutdown cooling mode of operation, which bound the EPU suppression pool temperature, so the stress analysis results are not changed. The highest stresses for piping and supports in the CS system are summarized in the table below, which indicate the loads are within code allowable values, although very close to the limits. Therefore, the associated piping and supports are structurally adequate for the EPU conditions.

L-MT-09-044 Non Proprietary Page 25 of 46 Core Spray Maximum EPU Results (Highest Interaction Ratio)

Maximum Pipe Stresses Pre-EPU EPU % EPU Load Condition IR Increase IR TH 0.91 8.9 0.99 Maximum Support Loads, Support TWH-86, Core Spray Pump Discharge Line Support Pre-EPU EPU Pre-EPU EPU Load Condition Load, lb Load, lb IR IR DW+TH+OBE 1471 1480 0.99 0.996 EMCB RAI No. 16 The PUSAR indicates that the MS piping pressures and temperatures are not affected by EPU. Please confirm that the main steam piping has no temperature and pressure increases due to the EPU and whether that includes main steam branch piping inside and outside containment including the main steam turbine bypass piping.

NSPM Response There are no temperature and pressure increases due to the EPU for main steam piping and its branch piping inside and outside containment including the main steam turbine bypass piping.

L-MT-09-044 Non Proprietary Page 26 of 46 EMCB RAI NO. 17 Steam flow and feedwater flow will increase as a result of the CPPU implementation.

The load due to the TSV fast closure transient is used in the design of the MS piping system. Page 2-31 states that Due to the magnitude of the TSVC transient load increase [at EPU], the transient event was reanalyzed. The main steam piping was then reanalyzed using this revised load definition.

a) Provide a quantitative summary of the MS and associated piping system evaluation (inside and outside containment), including pipe supports, that contains the increased loading associated with the TSV closure transient at EPU conditions, along with a comparison to the code allowable limits. For piping, include maximum stresses and data at critical locations (i.e. nozzles, penetrations, etc), including fatigue evaluation CUFs, where applicable. For pipe supports, state the method of evaluation for EPU conditions and confirm that the supports on affected piping systems have been evaluated and shown to remain structurally adequate to perform their intended design functions. For non-conforming piping and pipe supports, provide a summary of the modifications required to ensure that piping and pipe supports are structurally adequate to perform their intended design functions and the schedule for completion of these modifications.

b) For FW and condensate, please respond as in part (a) of this RAI.

NSPM RESPONSE Response to Part a The Main Steam system piping analysis results, including TSV closure loads are summarized below. The piping system was evaluated (by re-analysis versus scaling) using requirements from the existing code of record. The supports in the Main Steam piping remain adequate to perform their intended design functions. An updated status for PUSAR Table 2.2-2d is provided in response to RAI 12, Part b above. There are no non-conforming pipes or supports requiring modifications on the main steam system.

L-MT-09-044 Non Proprietary Page 27 of 46 Main Steam Inside Containment Maximum EPU Results (Highest Interaction Ratio):

Maximum Pipe Stresses Load Service Combination Level Node Stress Allowable Ratio psi psi S/Allow P+DW B 161 7709 15000 0.51 TH Range B 203 22940 22998 1.00 P+DW+OBE B U08 17823 18000 0.99 DW+TSV+SRV+SSE D U08 31261 36000 0.87 Note: 1. High Energy Line Breaks locations are not postulated inside containment.

2. Due to the revised analysis of the turbine stop valve closure loads, comparison to pre-EPU values is not meaningful.

Maximum SRV Flange Loads Inlet Flange Service Node Moment Allowable Ratio Load Condition Level ft-lb ft-lb M/Allow DW + TH B U07 14558 34083 0.427 DW + TH + Level B Dynamic B U07 39362 68167 0.577 DW + TH + Level D Dynamic D U07 65909 99750 0.661 Outlet Flange Service Node Moment Allowable Ratio Load Condition Level ft-lb ft-lb M/Allow DW + TH B U08 13663 31000 0.441 DW + TH + Level B Dynamic B U08 34907 62083 0.562 DW + TH + Level D Dynamic D U08 57547 91250 0.631 Maximum RPV Nozzle Loads RPV Nozzle N-3D Service Node Fx Fy Fz Mx My Mz Loads Level lb lb lb ft-lb ft-lb ft-lb Maximum Loads B 101 6667 18555 4979 67422 18193 98764 Allowables B 101 19392 51712 19392 258562 32320 258562 Maximum/Allowable B 101 0.344 0.359 0.257 0.261 0.563 0.382 Maximum Flue Head Anchor Loads Penetrations X7A, X7B, X7C, X7D - Side Bolt Evaluation Service Node Tension Shear T allow S allow IR Load Condition Level lb lb lb lb T/Ta+S/Sa DW+TH+SSE+BREAK (X7D) D 22 106702 17509 157500 96250 0.859 DW+TH+SSE+BREAK (X7A) D 30 107227 16683 157500 96250 0.854

L-MT-09-044 Non Proprietary Page 28 of 46 Maximum Support Loads MS Relief Valve Discharge Line Support RV25A-H1 (spring hanger)

Servic Nod Max Allowabl Min Allowabl e e Load e IR Load e IR Max/Allo Allow/Mi Load Condition Level lb lb w lb lb n DW+TH+

SRSS(TSV,SRV,OBE

) B 285 1341 1344 0.998 1162 780 0.671 Main Steam Outside Containment Maximum EPU Results (Highest Interaction Ratio):

Maximum Pipe Stresses Load Service Node Stress Allowable Ratio Combination Level psi psi S/Allow P+DW B X7A 6877 15000 0.46 TH Range B TURB 19441 22500 0.86 P+DW+TSV B 268 12236 18000 0.68 DW+TSV+SRV+SSE D 268 13795 26325 0.52 HELB DW+TH+OBE B TURB 27559 30000 0.92 Maximum Turbine Loads Load Service Node Mx Allowable Ratio Mz Allowable Ratio Combination Level ft-lb ft-lb Mx/Allow ft-lb ft-lb Mz/Allow DW B

• 32244 413000 0.078 171446 722000 0.237 DW + TH B

• 271321 413000 0.657 302310 722000 0.419

• Note: Loads from all turbine nodes were combined Maximum Support Loads Main Steam Line Support PS-16, Node 283 Max Service Load Allowable IR Load Condition Level Component lb lb Max/Allow DW+TH+SRSS(TSV,SRV,OBE) B Anchor bolt 20026 20731 0.966 Response to Part b The maximum Feedwater system operating temperature is 397.7oF at EPU conditions for the Feedwater piping from the outboard containment isolation valve to the containment and inside containment. This value is bounded by the original analysis temperature of 400oF. The design pressure for this portion of the Feedwater system is

L-MT-09-044 Non Proprietary Page 29 of 46 unchanged by EPU. Therefore this piping is unaffected by EPU relative to HELB postulation. The current design basis for Feedwater piping analysis does not include fluid transient analysis. The stress analyses for the Feedwater piping from the outboard containment isolation valve to the containment and inside containment are therefore unaffected by EPU.

The feedwater piping and condensate piping from the condensate pump suction to the containment isolation valves will be re-analyzed during the Feedwater and Condensate system modifications (reference response to RAI 7).

EMCB RAI No. 18 In accordance with Section 2.2.2 of the PUSAR, the main steam and associated piping system structural evaluation was performed to justify the operation of these systems at EPU conditions. This evaluation showed that one small bore branch line did not meet the displacement criteria. PUSAR further states that, "Additional detailed analysis will be performed to qualify this line or the piping modified prior to operation at EPU conditions."

a) Provide identification of the small bore branch line (size, system, location, function).

b) Describe the required displacement limits and their bases.

c) Since this piping analysis, with potential piping and or support modifications, is required for EPU, please discuss the reasoning for not including this information in your application. Also, indicate when necessary modifications, as needed, will be completed.

NSPM RESPONSE a) The branch line is a 1 inch instrument sensing line located inside the primary containment. The line connects one of the differential pressure sensing ports on the D steam line flow restrictor to a containment instrument piping penetration. This line is used for flow sensing in main steam line D and serves a safety related input function to the high flow Group 1 Containment Isolation logic that will automatically isolate the MSIVs in the event of a main steam line break.

b) A differential displacement of 1/16 inch for branch connection points was used as screening criteria in the piping analysis. Those in excess of 1/16 inch were noted as outliers needing further evaluation. The basis for the 1/16 inch criteria is:

L-MT-09-044 Non Proprietary Page 30 of 46

1. The 1/16 inch displacement produces an insignificant stress in the branch line which is typically supported by a standard deadweight span (span length from run pipe nozzle connection to first support on the branch).

2. Typical industry practice is to design supports with a gap of 1/8 inch to 1/16 inch.

Therefore the displacement due to EPU is absorbed by the support gap and produces minimal stress in the branch line.

3. The 1/16 inch displacement from the run pipe is considered a secondary stress since it is a deflection limiting stress. The piping system allowables for secondary stresses have significant margins beyond the code requirements especially when fatigue cycles are considered.

4. Typical industry practice is to evaluate main pipe run displacements much higher than 1/16 inch. Therefore the relative increase in stresses due to the EPU 1/16 inch increase will not be significant for the branch line.

c) The depth of information provided in the application was developed as described in Section 1 of the PUSAR.

To complete the evaluation of the instrument line noted above, a field verification of the distance between the pipe tap and first support inside the primary containment was required. This verification was completed during the current refuel outage. This small bore branch line meets code allowables, no modification is necessary.

L-MT-09-044 Non Proprietary Page 31 of 46 EMCB RAI No. 19 Page 2-31 of the PUSAR states that, SRV discharge loads are not affected by EPU.

Please clearly present your evaluation of the effects of the safety relief valve (SRV) discharge line and containment loads at EPU conditions, which demonstrates that the current design basis for containment dynamic load definitions for the SRVs are still valid and bound the EPU conditions.

NSPM Response The evaluation of the effects of containment loads at EPU conditions is presented in PUSAR Section 2.6.1.2. The containment dynamic loads include Loss-of-Coolant Accident (LOCA) loads and SRV loads. The evaluation of the effects of LOCA loads at EPU conditions is presented in PUSAR Section 2.6.1.2.1, and the evaluation of the effects of SRV discharge line loads at EPU conditions is presented in PUSAR Section 2.6.1.2.2. The conclusions of these evaluations are summarized here.

The LOCA dynamic loads include pool swell (PS), condensation oscillation (CO), and chugging (CH). Vent thrust loads, unique to Mark I containment types, are included in the evaluation. The short-term containment response at EPU conditions remain within the range of test conditions used to define the original PS and CO load definitions for Monticello. Vent thrust loads calculated with the short-term containment response at EPU conditions also remain bounded by the plant-specific vent thrust loads calculated during the Mark I Containment Long-Term Program (LTP). The long-term containment response at EPU conditions when chugging would occur are also bounded by the containment conditions used to define the original chugging loads for Monticello.

Therefore the current LOCA load definition remains bounding and applicable for Monticello at EPU conditions.

The SRV dynamic loads are influenced by changes in SRV opening setpoint pressure, the mass (length) of SRV discharge line (SRVDL) and suppression pool geometry, including the mass (length) of water in the discharge line at the time of SRV opening.

Since the SRV opening setpoint pressure and the SRVDL and pool geometry are not changing for EPU, the SRV dynamic loads for initial SRV actuation are not increased for EPU. The load definition for subsequent SRV actuations is not affected because SRV low-low set logic has been incorporated at Monticello to ensure that subsequent actuations occur only after the water level in the SRVDL has returned to normal.

Therefore the current SRV load definition remains bounding and applicable for Monticello at EPU conditions.

L-MT-09-044 Non Proprietary Page 32 of 46 EMCB RAI No. 20 Page 2-33 states that:

FW piping from the MOVs [downstream from the high pressure heaters] to the condensate pumps will be modified as a result of the replacement of the feedwater and condensate pumps, and will be qualified for full EPU operation as part of the modification. The current piping and associated components are adequate for operation within the capability of the existing feedwater and condensate pumps.

Page 2-61 indicates that:

BOP FW from the condensate pumps to the first isolation valves (IV) (outside containment) will be analyzed and qualified with the FW and Condensate pump modifications prior to operation at EPU conditions.

a) In addition to the minimum flow line modifications for EPU FW and condensate pumps (identified in Enclosure 8, Planned Modifications), what other piping modifications are anticipated?

b) Indicate whether piping (including supports) analysis at the EPU conditions of the above mentioned FW and condensate piping modifications (including minimum flow lines) has been completed and discuss the analysis results.

c) Provide an explanation whether any transients are applicable in the sections of piping mentioned above (including pump min flow lines) and evaluate their affects with regard to structural integrity of the proposed modifications of piping, pipe components and pipe supports.

NSPM RESPONSE a) Details of the modifications to the condensate and FW system are not yet finalized.

The design goal is to maintain stresses in the existing condensate and FW piping within code allowable limits of ANSI-B31.1-1977, including Winter 1978 Addenda (reference response to RAI 7) b) The piping analysis for the FW and condensate piping modifications has not been completed.

c) There are no fluid transients applicable to this piping. The piping is non-safety related/seismic Class II piping. It is analyzed for deadweight, pressure and thermal stresses. A portion of the piping from the 13A &B heaters to the FW pumps and from the FW pumps to the 15A&B heaters is analyzed to Class I

L-MT-09-044 Non Proprietary Page 33 of 46 seismic requirements. These stresses are imposed so that a Class II pipe rupture need not be postulated. (Reference response to RAI 7)

EMCB RAI No. 21 PUSAR, on page 2-33, to makes the following statement with regard to FW pipe stress evaluation:

A review of the small increases in pressure, temperature and flow associated with EPU indicates that the EPU temperature, pressure and flow conditions are bounded by the existing analyses. The original design analyses have sufficient design margin between calculated stresses and ANSI-B31.1-1977, including Winter 1978 Addenda Code allowable limits to justify operation at EPU conditions.

Explain the small increases in FW flow between OLTP and EPU and between CLTP and EPU that are bounded by the existing analyses, and whether the existing analyses contain flow induced loads at the OLTP or CLTP.

NSPM RESPONSE The portion of the Feedwater system evaluated in the PUSAR is from the motor operated (MO) valve downstream of high pressure heaters through the containment to the RPV. In this portion of the system, Feedwater flow increases from 7.235E+6 lbm/hr at CLTP to 8.497E+6 lbm/hr at EPU. The temperature and pressure used in the CLTP stress analyses bound EPU conditions. The current licensing basis (refer to USAR Sections 12.2.1.10 and 12.2.2) does not include flow induced load analyses for the Feedwater piping and none was added for EPU. The remainder of the Feedwater piping and Condensate piping includes more significant changes and will be evaluated during the modification design process of the Feedwater and Condensate pump and heater replacement modifications (reference response to RAI 7).

L-MT-09-044 Non Proprietary Page 34 of 46 EMCB RAI No. 22 PUSAR, on page 2-33, makes the following statement with regards to the FW pipe support evaluation:

The FW system was evaluated for the effects of seismic, deadweight and thermal expansion displacements on the piping snubbers, hangers, and struts. A review of the increases in temperature and FW flow associated with EPU indicates that the EPU conditions are bounded by the existing analyses.

Provide a discussion which shows that the FW flow induced loads on pipe supports in the existing analysis bound the EPU flow induced loads.

NSPM RESPONSE The portion of the Feedwater system evaluated in the PUSAR is from the motor operated (MO) valve downstream of high pressure heaters through the containment to the RPV (reference response to RAI 21). The design and licensing basis (refer to USAR Sections 12.2.1.10 and 12.2.2) for Feedwater piping analysis do not include consideration of flow induced transient loads. No flow induced transient analysis was performed for Feedwater piping. Flow induced vibration is evaluated and will be monitored during power ascension as discussed in EPU LAR Enclosures 9 and 10. The start up and power ascension vibration monitoring program will demonstrate that steady state flow induced vibration at EPU conditions remains within pre-established acceptance limits.

L-MT-09-044 Non Proprietary Page 35 of 46 EMCB RAI No. 23 a) Discuss whether there is any piping analysis, in the current design basis of the plant, that contains stratification or discuss whether there is any CLTP stratification monitoring currently in place.

b) If a stratification phenomenon currently exists, explain how these stratification locations have been evaluated at EPU conditions and provide a summary of their evaluation results.

NSPM RESPONSE MNGP piping analyses do not include consideration of thermal stratification. To validate this, MNGP installed thermocouples on top and bottom of the horizontal Feedwater lines at elevation 952-10 in the drywell and monitored conditions during startup and shutdown to verify that no global thermal stratification occurs on Feedwater lines due to interaction with RWCU. This was monitored over several start ups and shutdowns with very similar results each time. The results did not show any sign of significant stratification. With relatively minor changes in temperatures in the Feedwater and RWCU systems, it is expected that no global thermal stratification will occur in these lines at EPU.

L-MT-09-044 Non Proprietary Page 36 of 46 EMCB RAI No. 24 Consider the two statements below:

LAR Enclosure 10, on page 3 of 16, states the following:

If the vibration level in the main piping in these systems [(FW and MS)] is greater than 50% of the acceptance criteria, then an engineering evaluation of the small bore piping will be performed to ensure that the steady state stresses are within the endurance limit.

In response to NRC staff RAI, CLTR for the EPU generic evaluation states that:

[T]ypically the measured piping vibration levels of the MS and FW piping are only a few percent of [the acceptance] criteria. Hence, the vibration levels of the large bore piping are small and therefore the vibration levels of components and branch piping attached to the large bore piping are not of concern. However, if during testing, the vibration levels of the large bore MS and FW piping are found to be significant, ((say 50% or higher of their acceptance criteria,{3})) then the attached components and branch piping connections will have a higher probability of fatigue failure relative to operation at the original power level. Hence when the measured MSL or FW large bore piping vibration levels reach ((50% of{3})) their acceptance criteria, the attached branch piping connections will be further evaluated.

EMCB RAI 24 a)

Please revise the statement of Enclosure 10 of the LAR to be in accordance with the generic CLTR evaluation, in that if the vibration levels of the main piping reach 50% or higher, an engineering evaluation of all attached branch piping, not just for the small bore, will be performed to ensure that the steady state stresses are within the endurance limits. As this was the intention of the CLTR statement.

NSPM RESPONSE a) The subject statement of Enclosure 10 is revised as follows.

If the tested vibration level in the main piping in these systems (FW and MS) is greater than 50% of the acceptance criteria, then an engineering evaluation of the attached branch piping connections will be performed to ensure that the steady state stresses are within the endurance limits.

L-MT-09-044 Non Proprietary Page 37 of 46 EMCB RAI 24 b)

It appears that the 50% was based on the CLTR statement that, measured piping vibration levels of the MS and FW piping are only a few percent of [the acceptance]

criteria. However, in the Monticello case, from readings taken at 100% CLTP, vibration resulted in levels well above just a few percent of the acceptance criteria. At CLTP, 10 locations came in at above 20% of the acceptance criteria for FW and MS. Inside containment, the maximums were 14% and 32% of the acceptance criteria for FW and MS, respectively. Outside containment, the maximums were 43% and 34% of the acceptance criteria for FW and MS, respectively. Using the EPU expected vibration increase of 32%, the CLTP values of 14, 20, 32, 34 and 43 percent of the acceptance criteria are projected for the EPU to be 18, 26, 42, 45 and 57 percent of the acceptance criteria, respectively.

EMCB RAI 24 (b) 1

.Using the 50% or higher criterion, one location has been predicted to be 57% of the acceptance criterion. Please discuss whether evaluations have been performed for branch lines in the vicinity of this location? Provide a discussion of the evaluation results.

NSPM RESPONSE An evaluation of the branch lines in the vicinity of this location has not been performed.

The 57% projection is conservatively determined, such that the actual testing results may not exceed the 50% criterion. The piping located outside containment was designed to ASME B31.1, yet the acceptance criteria was developed using a conservative approach. If the actual results do exceed the 50% criterion, the response to part a) will apply.

EMCB RAI 24 (b) 2 Provide a basis for justification that the 50% criterion, which the CLTR recommends for cases where piping vibration levels are only a few percent of the acceptance criteria, is applicable for Monticello, where the vibration levels even at CLTP have reached well above a few percent of the acceptance criteria, as shown above.

NSPM RESPONSE The acceptance criteria were conservatively developed so that even if some of the vibration levels at CLTP are more than a few percent of the acceptance criteria it is not an indication that the piping may have a fatigue failure of a branch line.

L-MT-09-044 Non Proprietary Page 38 of 46 The piping vibration levels that have reached at least 30% of the acceptance criteria are mainly located in the steam tunnel and turbine building. This piping is designed to ASME B31.1. The seismic analysis used to determine/locate the supports for these piping systems is typically a static analysis. The acceptance criteria for the MS and FW piping were developed by conservatively applying flat dynamic spectra in all three orthogonal directions. The flat dynamic spectra used to develop the acceptance criteria are extremely conservative since the same magnitude is applied to all frequencies. The steam tunnel and turbine building piping do not have many supports. Due to the low number of supports, there are numerous low natural frequencies that normally would not be subjected to the same magnitude of input than would be at higher natural frequencies.

The input used to develop the acceptance criteria is very conservative because the acceptance criteria were developed prior to collecting any steady state vibration data. If the steam tunnel and turbine building acceptance criteria were revised to reflect more realistic magnitudes, the margin would be greater.

L-MT-09-044 Non Proprietary Page 39 of 46 EMCB RAI No. 25 With regard to the reactor pressure vessel (RPV) evaluation for EPU, Page 2-45 of the PUSAR states that:

The Top Head and Cylindrical Shell and the Stabilizer Bracket were not evaluated for fatigue at the time that the OLTP evaluation was performed, and have not been evaluated for EPU.

Monticello USAR Rev 24, Section 4.2.1 states that:

[T]he reactor vessel was also designed for the transients which could occur during the design [ ] life. The reactor vessel was analyzed for the cycles listed in Table 4.2-1.

Provide an evaluation which shows that the RPV top head and cylindrical shell and the RPV supports will be structurally adequate at EPU conditions for the renewed plant life.

NSPM RESPONSE The purchase specification for the Monticello RPV defines the regions and components of the vessel that are to be analyzed. These regions/components include the head closure, bottom head, shell adjacent to the reactor core, reactor vessel supports and stabilizers, supports for reactor vessel internals, control rod drive penetration, feedwater nozzle, poison nozzle, emergency core cooling nozzles, drive system return nozzle, and all nozzles 10 or larger in size.

A summary stress report was generated at the time of vessel fabrication for the vessel shell and top head. ((

))

The summary report for the shell and top head includes a summary of all major discontinuities in the shell and top head. These include the dryer hold down bracket, guide rod bracket, steam outlet nozzle, steam dryer support bracket, stabilizer bracket, feedwater bracket, feedwater nozzle, core spray bracket, core spray nozzle, top and bottom insulation brackets, recirculation inlet and outlet nozzles, shroud support, support skirt and knuckle, and the jet pump riser pad. In addition, refueling bellows reactions were considered.

Duty cycles specific to the shell and top head are not defined in the purchase specification; ((

L-MT-09-044 Non Proprietary Page 40 of 46

)) the shell and top head are considered structurally adequate for operation at EPU operating conditions for the license period of 60 years.

EMCB RAI No 26 Table 2.2-4 of the PUSAR shows that the fatigue CUFs for the recirculation (RRS) inlet nozzle (Ri) and FW Nozzle significantly increased for EPU by 146% for Ri and 47% for FW nozzle, placing the FW nozzle within approx 8.6% of its limit. Provide an explanation for these significant EPU CUFs increases, confirm that these CUFs are to the end of renewal life and assure that all required transients at EPU conditions have been properly included for these fatigue evaluations.

NSPM RESPONSE An extensive review of the component history was performed for these components.

The key items of the review that resulted in the CUF changes due to EPU are shown below.

((

))

Legend: 3Sm = 3 times Sm, where Sm = Design Stress Intensity

L-MT-09-044 Non Proprietary Page 41 of 46

))

Legend: Sn =maximum primary plus secondary stress intensity Kt = stress concentration factor Ke = elastic-plastic stress concentration factor Salt = amplitude (half-range) of stress fluctuation N = number of allowable cycles n = number of required cycles u = usage factor for the given stress U = cumulative usage factor

((

))

Legend: Sn =maximum primary plus secondary stress intensity Fth = thermal peak stress Kt = stress concentration factor Sp = peak stress Salt = amplitude (half-range) of stress fluctuation N = number of allowable cycles n = number of required cycles U = cumulative usage factor

((

L-MT-09-044 Non Proprietary Page 42 of 46

))

Legend: Sn =maximum primary plus secondary stress intensity Fth = thermal peak stress Kt = stress concentration factor Sp = peak stress Salt = amplitude (half-range) of stress fluctuation N = number of allowable cycles n = number of required cycles U = cumulative usage factor

L-MT-09-044 Non Proprietary Page 43 of 46 EMCB RAI No. 27 In Table 2.2-9 of the PUSAR, some of the locations are shown with --. Please explain what is meant by this designation.

NSPM RESPONSE The '--' designation is used in the Location column under the EPU portion of the Table and means not applicable for Items 8 and 16. For Items 6a, 6b, 7, 12 14, and 15, the '--'

designation means the specified locations of the stresses for the components in question are not known.

EMCB RAI No. 28 In Section 2.2.2 of the PUSAR, it is stated that, The effects of FIV induced stresses at EPU conditions on safety-related thermowells in the MS and FW system and the sample probe in the FW system were evaluated and indicates that they remain acceptable under EPU conditions (see page 2-28 of the PUSAR). However, Enclosure 8, page 5 of 9 states, Replace or remove the thermowells in main steam piping to insure appropriate margin for flow induced vibration. Provide a quantitative summary of the evaluation that supports the acceptability of the thermowells and sample probes in the MS, FW and related piping systems. Identify nonconforming component(s) and provide description of their modification(s).

NSPM RESPONSE A quantitative summary of the evaluation results are provided in the table below:

Component CLTP EPU Zero-to-peak fs/fn Reduced Zero-to- fs/fn Reduced stress (psi) peak stress Velocity (psi) Velocity FW Thermowells 2683 0.57 1.30 4536 0.67 1.53 FW Sample 1627 0.31 0.70 2332 0.36 0.82 Probes MS Thermowells 1308 0.70 2.32 2809 0.82 2.73 The stress results were compared to the 13,600 psi endurance limit for all the materials of the probes and thermowells. At EPU conditions, all of the stress values are below this endurance limit and thus the thermowells and sample probes are structurally adequate.

However, it is desired to reduce the ratio of the vortex shedding frequency to the natural frequency of the MS thermowells (TE 2-127A & B) to the CLTP value to minimize the

L-MT-09-044 Non Proprietary Page 44 of 46 potential of the system jumping into resonance. Reducing the length of the thermowells by 10% will accomplish this goal. Currently two options are being evaluated; either replace the thermowells with shorter ones or remove them altogether. Final resolution of this issue is now scheduled for the 2011 refuel outage.

EMCB RAI No. 29 Page 2-59 of the PUSAR states that:

The temperatures, accident radiation level, and the normal radiation level increase due to EPU. These effects are not considered to have an adverse effect on the functional capability of nonmetallic components in the mechanical equipment both inside and outside containment.

Please provide a justification that the radiation due to the EPU is not higher than the radiation damage threshold of the non-metallic parts of the resilient seated check valves, hydraulic snubbers and flex joint bellows affected by the EPU.

NSPM RESPONSE The predicted dose increase due to EPU operation was determined for all plant general areas. The prediction is that the dose will increase slightly. MNGP will perform plant radiation surveys during power ascension testing and at EPU (power operation and post shutdown) to confirm predicted radiation dose rates.

MNGP has active and formal programs in place to properly manage the slight increase in radiation expected for EPU. The subject components are procured and designed for the applicable service environments in accordance the requirements of the Quality Assurance Program. This program includes requirements to assure that plant equipment is suitable for the intended service, and is of acceptable quality consistent with their effect on safety.

The MNGP Check Valve Program closely monitors valve reliability. The program monitors check valve maintenance history and check valve failures. The check valves with non-metallic seals are included in the program. Valves with non-metallic seats receive regular maintenance including inspection and bench testing. The valves are functionally evaluated during maintenance and replaced if necessary. The O-rings and seals are typically replaced regardless of condition. This program has provided reliable check valve performance to date at Monticello, and the slight increase in radiation due to EPU is not expected to have an adverse effect on continued reliability.

Like the check valve program, the MNGP Snubber Program closely monitors snubber

L-MT-09-044 Non Proprietary Page 45 of 46 reliability. The program monitors maintenance history and snubber failures. The in-service requirements are delineated by Section 3.4.3 of MNGP Technical Requirements Manual. The installation and maintenance records for each safety-related snubber are reviewed at least once every 24 months to verify that the indicated service life will not be exceeded prior to the next scheduled service life review. The service life of a snubber is evaluated via manufacturer input and thorough consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replacement, spring replacement, in high radiation, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records provide statistical bases for future consideration of snubber service life. In addition, to address seal failures specifically, Monticello has assigned a maximum service life of 10 years for all hydraulic snubbers regardless of installed location in the plant. If degradation or damage is detected, the number overhauled will be adjusted. This was noted in an NSP letter (L. O. Mayer) to the AEC (J. F. OLeary) dated October 1, 1974.

This program has provided reliable snubber performance to date at Monticello, and the slight increase in radiation due to EPU is not expected to have an adverse effect on continued reliability.

A database search of the MNGP plant equipment did not identify expansion bellows with elastomer components. Regarding expansion joints, the plant systems with rubber expansion joints were identified as part of the MNGP License Renewal Program. These components are not located in safety related systems (e.g. Condenser, Service Water System). The program determined that changes to material properties for rubber required a source strength of 10E7 Rads, and concluded that these components were not susceptible to hardening and loss of strength caused by radiation as there is significant margin to this value. This margin exceeds that which may occur due to the conservative 13% increase in radiation expected for EPU. In addition, the systems that were identified to contain rubber expansion joints are within the scope of the MNGP Maintenance Rule Program. The program monitors system reliability and a significant increase in failures of rubber expansion joints for these systems has not been noted.

The slight increase in radiation for EPU is not expected to have an adverse effect on the reliability of systems containing rubber expansion joints.

L-MT-09-044 Non Proprietary Page 46 of 46 EMCB RAI No. 30 Page 2-59 of the PUSAR states that:

The Monticello design and licensing bases do not require a formal mechanical EQ program like the EQ program applied to electrical equipment.

What program used at Monticello establishes the capability of active safety-related mechanical equipment and their components to perform their required safety function for the life of the plant during postulated normal and accident conditions?

NSPM RESPONSE Monticello does not have a formal mechanical EQ program. The remainder of the PUSAR paragraph cited above describes the programs that are in place at Monticello.

The key elements are design control, testing/preventive maintenance and equipment monitoring in accordance with the maintenance rule. A key element of the maintenance rule is to also incorporate industry-wide operating experience into the program. The integrated effect of these elements provides reasonable assurance that important systems, structures and components will be capable of fulfilling their intended functions.

ENCLOSURE 4 United States Atomic Energy Commission - Safety Evaluation by the Directorate of Licensing, Docket No. 50-263, Monticello Nuclear Generating Plant - Analysis of the Consequences of High Energy Piping Failures Outside Containment, July 29, 1974

ib- 5&263 6;

TO= &at&eahtr& *0li?:meigf% 7, 1973, B + ~ L M ~ a repert orp Pwtulated oar l e t ~ r e a r 18, 1972.

$8, B B , a d Jam 16, 1973, con-

~ h w c $ne aergy f l u i d pQ$x?g e&srH* t w m to be avraluated to assure m q e p 8 m 8 skw ebsarw 5 , 1973, a m e t i n g w a s held t o tbm d , r a x h m d the resdts sf your prelidoary study 0% ehe : m & i e & 2 s iiw-r Wef1zLveg PHsaamt, b p ~ nconapletron of our

&=LEA& r n * ~06 yumzb s p m I ~ B ; E ~ ePh~r & X % & w t bofn the analysis w m q m t & 2s. ow 2ee-r ~f J a w q 8 8 , 1974. Your response to bekeg g ~ Y ~~Z & ds, ass&.

0 s re&- @? ;:-a h E o ~ m t i o ayou have s d d t t e d on the S=3em W e a E w P h f , we %awe d e t e s d n e d thac a f t e r c o w 1  %%aac&ow,zbe plmt w o d d w2tbstand the con-I I

m t m e s k t $jigs energy fluid piping outside the 92 c k ce&iUrp ro initiate and aaintain the 0 . You have informed us that these mod-I f f Z a t & m s w e r e ~*-2~ted h d n g the 3 p d n g 3.974 refuelling outage. There-fare, %SE ern&& C?IS& cbe ;%at 0 2iuelear Generatin$ Plant now conrplies i e5e req& z.3 ~ E f e ox= Be r 18, 1972 fetter. A copy of our 1 relaced Safeey 6vaBsaz%oa-& eadosed.

i?Xar9P. Caller, Assistant Direceor for CQperatingPaactors DZrectorate of LfcensLng

,*o;hay Z. S ~ i s m maulre Srlbia, Rob- and Kessler l t l 2 K Street, 3. W

• UasMng,~on, 9. C. 261436 Dr. Ed PrytPm Mfanessea PoISution Cmtrol Agency 3939 B. C o w r y &ad 82 Roaeville , Nimesora 55114 xr- Gary Villi-Federal dctlvit i e s Branch

& v & s ~ w t a z PPsssecdon Agency 1 X. @a&er Dsfre, Etosar 822 Blicago, HjLlhois 60606 e a Bzugam, B5zacZaar St. Palap Fo2lutlonr GmCroP Sem5ces l;BB% %-E Sfret St. P a d , esoca 551G2

r BI, X9m, d J a w r y 26, 1973, the A t o a c rtabf m e f a m d ~ ern w%ebmStates P ~ x rn s d s m w t b t e ghat the d PBmc is adamat+ to withstrlnd the effects bf& eerw FPuM p l p b g system ciiltside b s t2~edouble-cded rupture of the a ~ fee*ater d s y s t m . It w a ~furtiter f z5e evadwt%m indlcateb that changes assre sage g h a t shurdown, information h -diElea?tiena; w o d d be required.

CrAeer%a &B k f h h e v d w t P m were 1.1tc.luded 5n t h e letters.

B mee- OP= held on F&rwq 5 , h973, to d-isccuss the information a r e & p m l f d l e oa *e & n ~ e c U o Plane d e s i g n concerning postulated

-uw &re er",eri,a, a d to assess those areas where dm seqalr&. I n respome t o our letters, a mast@ hw~a eaergy pipe ruptures o u t s i d e containment w m b i l d by %reem Smees Pwer C w i t h letter dated S e ~ e e e r7s 1973. A ~ W Q ~ I BPeeker E frm Boar&satn S t a t e s Power Company dated

%x& 8 , 1974, a m e x & aeBd.itars1 ~ s & s t i o n o -in a Ietrer fraaa the staff at& 3m-q 18, 397.4.

ry OE the ~ r Z e ~ r 3 4. a requ5.r-mes included in our letter of r 18, 1972, 2s set: Porch bdm-:

. haewzm G: api t a d sereeeww laeceesary to :sbaucdm mba.ala i6t: So a saf@ shut&wn c~flxiit%oa, a ~ a m ma t& mr&ased sing3e eet&vc CaUure of pmewtd ~ Z - Q C , s h M d be p16vdded ffraa all effects resulting fm ~ q k ~ r n 2as p f w earq l-g lkw srsergy fXuid, ~ k r the e

s ~ m a e o r e$@ad prm-8 ~m::A4mssf the Nuid w e e d 200°F w d 275 po&* mwwt2wXy, up ~ E Oas4 indudlng a double-ended wtwa & a- pQ%s. &n&s e h ~ d dBe msuzrted to occur in those e "pip@whip e~iteria". The ryture stffeeks m be Wee pipe whip, struetmd \ i n c l u d W a k a eEEee&ksef $@g: ~ i ~ m t a dl emrixo~aiwtaf.

b e '%w agbtaftba, p f e . P m o f a p 9 p n t and structures necessary t o the reaesdcba: In i r 2n e safe s h u t d m condition, b t a d single active failure of protected e the enviroxmwtzkl and structural jet laapfn@?w%It) resulting from a shgls ope% etbatk at the w s t odvewse Location i n pipes carrying f l d d mmtd im t&a i t y of thL equipr~ent. The a f z e of the d g b a B d 4 be as to be I12 the pipe diameter in length and I P a ~ ~ s a nt m h e ~ UP.

mties ~zR&&& &s folloring p5ping s y s t w containing High Emre flw2&:

Winiw, F&tkact%BP~ih, a d AwxjiPiary St@& System

&e&s?atier SyraarelB eEPdmb:e s y a a

&sctasr Core BsoB.&5w Coo1bg System IRCIC)

H4& Pressanre -bat 2njectt9o System (HE11 Reactor BPaEzr Clmmp System CXWCIS)

R e s i i a d &at &sov.8%1System <Rat)

S q l e L 5 . n ~dbviromentsll Effects Only)

-&ceas or S y s t w ~-Affected

~ Za)p H i ~ hEneteT Pipe Breaks ewdwrirw w a s c&uct& of the effects of hi& energy pipe breaks on &e foUm$. systew, wpnpwents, and structures, which would be

m e m q (&. -tIa @ & - t i a s , &pendw as the effects s f the b"s&fi M &at?-. mBh93n2, azdj mantain cold s h u t d m eenditions:

%C app2beme Bm v v M & rmdes of h.Ls

• ation of all m E q &at& %a- a% e br ations and evaluated mm aljl of tftfs inforraatiarn, e k EoUm* s p a f i e e- sf corncerez where &e potential es &&t be myere! here swePf-jtc corrective action w a d safe erolX&%bat of the pxmt.

-age .iimIudPmg, t b hale-ended rupture of the lawss p 3 p in a s y s t a , a d sm11 leakage cracks up t o the iB.es- basis ski? h v e bemi a ~ w M e r &for the =in steam tumel, t k m a h e b d l h g , the Z C S room, and c%e valve compartnents.

e d m s a failure of a ~ a i n steam line preslsasdze. P coqar-t io 1.4 p i g m x b v4~ntarea! tely 500 ft . The vent area is sd%ie%me eo 1prevmt d-ge or loss of safety equipment and to keep &e peak pxessure we11 be1aw the 8.4 p ~ i gd e s i g n .

Xn t b l a a h s the effects of a a n s t e m l i n e break

=ere mmZderee i s rbe des- eases. The resultant pressure was d d a t & to h r c ~ s to 3 12.2 p i g ,

& P a a m ef Cbs &weer F@e3&t;Bm S y s t m (RGZC) 2kse =&& 4.a the Ilms o f one emxgewy service w k e r 3

• a 1 +%s ib 6% s-]be dal3u-r'~d$s&led t h e xedmdant mm$tsrw~f conwetion @ a d d be assamplished w e e ~ t c ,rsw%re$e nmts in t h e t o en&le f &eh @me.

Ira t b B E $ m ~ a r t w a tcould result e a t m t i c I?rpolatZow is achiewd.

in the c ~ w r % a t : has been cal-Zeh b b%w the structural. capabilities A pe&atd 3*XZ?.nPbi* e a c X~3 . m Fa5-e 2x1 the cleanup system s resonlts i n a single-ended t

sf%.=& =all k9b-t- 3s ffakhved. A check valve in the a a z e e 2 m Pam &e feehater p p ~ i q l :would prevent extensive

& pressure ro?asudthg from a pipe failure t a e h a a e r r o o m , aari 0.2 psig i a the dwgg~rr.capacities of these compartments area 16.0 ~ s 5s: d 33th & W e e x e b w e r a d p w corPpartwnts.

r e ~ c o3 t 23, =-;rbbe be%ldZ;I;sareas were considered for the e g f e c t c-& +%%p a d jet iqbg-nt from the m3ins t e m feed-

~ ~atadt zc- ekeQ-te l h e s . The seeam trmnel has been designed wi* zbdck ~ e f z - s r c e deoaerete t a p b l e of withstanding large static e 1 0 s The reMore6td concrete s t e m tunnel in which stease a-- fe&mter lines are routed fxoln the primary nn~: &E t~urbioer c a i s subjected only t o the loads pxp- azC a l2we h a d f r o m the floor on top of tshe tunnel or feedwttter line in the main steam rapcare of the HIPCX, ECIC turbine steam i n l e t 2.5ne. E ~ ~ e v e rloss, of these ELms would not itzpai5.r safe! s & ~ E Z = of a e plant.

broke% frn&aeer ldna f a rbe areat i a i eers W e ) eeml&C ~ W C WBG loss EQ redundant Edlure of the aexmmia~n"loc4r. The pZp3q rmsrafaats 2x1 zh%s area will reduce Eke rPe~4sm&Be& $ w $ q a QS a h wazangnc flmr ant? ehereby

)PQtSlh, %b s, 1 1 : ~ pfgim w a d not: cause a, hazard rtcs safeguards ebe m8, Pwb1m ialeg: %.siseutd a"mve chc corus, therefore SWC - % O m & m d d C ~ W IfK wc Bt l w a l .

O a e r M@ -m ggRm sweh as tb kc hiaes &ad rea~forW ~ L C P Aims ere l w t & se& e b $ . ehebr rtaptasze w u l d not cause.

caa W &em. & *ip of excher t k RCIC or EPCI s t e m

%as& @h mms c q a t z ~ er o ~ l d& w e sysrenD i s s l a t i o n w3we e b or RCEC UWW. EI-ver, the rapslttznt damage w e d B mt Tr %fa W t d e m o f ehrg mn3t.

Tba m b e m t m 8 Ply iselated fro@a11 high energy equfp~ggntnor i t s voncf lation s ~ e m wPXl be afEd?iii=1%ed by av5rmentaP effeees caused by a P*kUre 0% a $Qh ernerw l i e .

e en%*~o-aeb-eB. effeers; &ieh mu14 be caused by the rupture cf a Mgh meagy l&@e, Mvessr? t ratuse, pressure, and humidity s *ich vzre us n the evaluation of safety iewed the l P c ~ s e eassessent of tar effects on safety related qd-~bs. EC f h d &hatsafety related equipment -has been designed

&a IWts %nexcess cf gostulared conditiom which corzld arise froa ehe m t w e of a WgIn eaergy Ilae.

Wifi=tf@w to zhe e x i s t k g facility are currently being undertaken by 3ertBem Smtes i a a eider to assure that the desiga - a i l 1 have adqmte safety m g 5 . z ~3.n the event of a high ene+R- f i r e rupture

-&side ~'aeeeat t. These nodificarions are to be f m p l e t e prior '10 r e s ~ a % d ~ &pel a t fcrllowing the curreris, refueling outage, tar-rentXy SC&&UI~& r0 em3 on or about Xay 15, 1976.

@b eke b b OI r b b rev&- O% W g n E o n t 2 ~ ns~bat~ibfigd to us and on

&&~moga%s wgtb Mrrzbm S l , r H e ~ Pwera we %dad z b e t k l r maese-nt m f &B c w q m e w @HI $?KG% ww m e E~ailurm0uts5de e s n t a i n m t bs oitcep~&8e. S e M L f 2 r n f - m are ~ b t s s a l y . Ue have eowluded

&be ~ e & m t mcweq-ew af E ~ S Cp t u & e & hQ$i energy p i p e Pii~;i$essm,fez as, wgll not prevent the capability r$;owMm?$oa t$m earnfarent da the single d&lure e as dmes&bed i r a our letter o f ma  %;l.easaab e s c w d &ae at d%%imcism were eoarpleted prior to m e m i = %o m r a g i m f m &ke ;%BP]~%Q&8974 reBuelh,.rg outage. With the

~-2ehLgpi$ ob d % f f ~ , o ~ , l 6Ebre

1) w f $ 'irmsonable assurance t h a t at B w l & dap:q sf the p%abP%e w&U be rtndangered by continued

~ r $ ~ t B o ~ .

.m

@grating Reactors Branch #2 Wectoraea of Licensing b

@peratringReactors Branch 02 DZrectorate of Licensing